1. Field of the Invention
The present invention relates to spacer grids for dual-cooled fuel rods, and more particularly to spacer grids for dual-cooled fuel rods, the spacer grids having upper and lower cross-wavy-shape dimples, capable of improving support stability of dual-cooled fuel rods.
2. Description of the Related Art
FIG. 1 is a schematic perspective view showing a conventional fuel rod assembly. FIG. 2 is a schematic top-down cross-sectional view showing the conventional fuel rod assembly. FIG. 3 is a schematic top-down view showing a part of a spacer grid applied to the conventional fuel rod assembly. FIG. 4 is a schematic perspective view showing a part of a spacer grid applied to the conventional fuel rod assembly. FIG. 5 is a schematic perspective view showing a unit grid strap of the spacer grid supporting the conventional fuel rod assembly.
As shown in the figures, the fuel rod assembly 10 is made up of fuel rods 11, guide thimbles 14, spacer grids 15, an upper end fitting 12, and a lower end fitting 13.
Here, each fuel rod 11 is configured so that cylindrical uranium sintered pellets are housed in a zirconium alloy cladding tube, and high-temperature heat is generated by a nuclear fission reaction of the uranium sintered pellets.
Meanwhile, each guide thimble 14 is used as the passage of a control rod that moves up and down to adjust output of a reactor core and to halt the nuclear fission reaction. Each spacer grid 15 is one of the components of the nuclear fuel assembly, and has a plurality of unit grid straps, each of which includes a spring 17 and dimples 18. The spring 17 and dimples 18 function to support the fuel rod 11 so as to be arranged at a designated position.
The upper end fitting 12 and the lower end fitting 13 function to fix and support the nuclear fuel assembly 10 to upper and lower structures of the reactor core. The lower end fitting 13 includes a filter (not shown) for filtering foreign materials floating in the inside of the reactor core.
The spacer grid 15 is typically formed of a zirconium alloy, and includes fuel rod cells in which the fuel rods 11 are supported and guide thimble cells into which the guide thimbles 14 are inserted. Each fuel rod 11 is typically supported in the fuel rod cell at a total of six support points by a total of two grid springs 17 located at two grid straps one by one and by a total of four dimples 18 that are located at the other two grid straps and above and below each grid spring two apiece.
Further, cylindrical uranium dioxide (UO2) pellets are charged into the fuel rod 11. A coolant rapidly flows from the bottom to the top of the reactor core through sub-channels 16, each of which is surrounded by four fuel rods 11 or by three fuel rods 11 and one guide thimble 14, in an axial direction. Here, the sub-channel 16 refers to a space surrounded by the fuel rods 11, and a passage that allows fluid to freely flow to neighboring sub-channels through an open gap between the fuel rods.
Meanwhile, when a spring force or a spring constant of the grid spring 17 located in a restricted space is too small, it is impossible to support the fuel rod 11 at a designated position, so that support soundness of the fuel rod 11 may be lost.
In contrast, when the spring force of the grid spring 17 is too great, defects such as a scratch may occur on a surface of the fuel rod 11 due to excessive frictional resistance when the fuel rod 11 is inserted into the fuel rod cell of the spacer grid 15, and it is impossible to properly cover lengthwise growth and thermal expansion of the fuel rod 11 caused by neutron irradiation when the reactor is being operated, so that the fuel rod 11 can become bowed.
When the fuel rod 11 is bowed, it comes near to or in contact with the neighboring fuel rod, thereby narrowing or blocking the coolant channel, i.e. the sub-channel 16, between the fuel rods. As a result, heat generated from the fuel rods is not effectively transmitted to the coolant, thereby causing the phenomenon of the temperature of the fuel rod being locally increased. Accordingly, there is a high possibility of generating a departure from nucleate boiling (DNB) that is a major cause of reduced output of nuclear fuel.
Further, the reactor coolant flowing around the fuel rods 11 is typically known to cause a flow with large turbulence, i.e. a flow of a high Reynolds number, in order to promote thermal performance. The turbulent flow of the coolant around the fuel rods becomes a major cause of flow-induced vibration of the fuel rods.
This flow-induced vibration of the fuel rods is responsible for a relative motion of the fuel rod 11 sliding on contact surfaces of spring structures of the spacer grid 15. For this reason, the fuel rod and the contact surface of the spring structure undergo local attrition. Thus, the fuel rod is gradually damaged, i.e. fretting damage is caused to the fuel rod.
In detail, a cladding tube of the fuel rod is formed of a thin zirconium alloy, and is charged with uranium dioxide pellets as fission materials and an inert gas under pressure. The cladding tube of the fuel rod is supported at multiple points by the contact force of the springs and dimples of the spacer grid.
The flow-induced vibration of the nuclear fuel caused by the high-speed flow of the coolant in the reactor core during its operation occurs over the life span of the nuclear fuel, so that the fretting damage caused by the relative motion occurs at the support points of the fuel rod.
When the fretting damage develops to the point that the cladding tube is penetrated, this leads to an accident wherein the fission products inside the fuel rod contaminate the primary cooling system of the nuclear power plant, and thus the operation of the nuclear power plant may be stopped, or an enormous repair expense may be required.
Meanwhile, as shown in FIGS. 6 and 7, a dual-cooled fuel rod 19 having an annular structure in place of the cylindrical fuel rod 11 is disclosed in U.S. Pat. No. 3,928,19b2, and U.S. Pat. No. 6,909,765.
Here, the dual-cooled fuel rod 19 having an annular structure is made up of sintered compacts or pellets 20 formed in an annular shape, an inner cladding tube 19a installed on inner circumferences of the pellets 20, and an outer cladding tube 19b installed on outer circumferences of the pellets 20. The coolant is allowed to flow to the inside and outside of the dual-cooled fuel rod 19 so as to perform heat transfer doubly, so that it is possible to maintain the inner surface of the dual-cooled fuel rod 19 at a low temperature and to obtain high burnup and output.
When the central temperature of the dual-cooled fuel rod 19 is kept low in this way, there is a reduced possibility of fuel damage caused by an increase in central temperature of the nuclear fuel, so that it is possible to increase the safety margin of the dual-cooled fuel rod 19.
However, to be structurally compatible with the core of an existing pressurized water reactor (PWR), it is impossible to change positions of the guide thimbles 14 in the nuclear fuel assembly 10′. Since an outer diameter of the dual-cooled fuel rod 19 is increased, a gap between the dual-cooled fuel rods 19 is considerably reduced, compared to a gap between the existing cylindrical fuel rods.
For example, when the nuclear fuel assembly is manufactured according to a candidate design draft for the dual-cooled fuel rods having a 12×12 array, the gap between the dual-cooled fuel rods is reduced from the existing 3.35 mm to 1.24 mm.
Thus, due to the narrow gap between the dual-cooled fuel rods, the spacer grid developed up to now cannot be used as a support structure for the dual-cooled fuel rods 19 unless it is changed.
That is, a thickness, 0.475 mm, of the unit grid strap of the existing spacer grid is subtracted from an interval, 1.24 mm, between the dual-cooled fuel rods, and the resulting value is divided by two, so that an interval between the unit grid strap and the dual-cooled fuel rod is about 0.383 mm. As such, it is impossible to design the spring having spring rigidity and hydraulic characteristics (mainly, pressure loss), both of which the existing support structure has, by applying a shape and a support position as in an existing leaf spring within this narrow interval.